The present invention broadly relates to a method of calculating failure event probability in fuel rods of nuclear reactors and, more particularly, to an apparatus for controlling the output power of the nuclear reactor equipped with a fuel rod soundness observation device which permits the control of operation of the nuclear reactor in accordance with the fuel rod failure event probability determined by the above-mentioned method.
During starting up of a nuclear reactor, the temperature of the fuel pellets in each fuel rod is raised so that the fuel pellets are thermally expanded to reduce the gap between the fuel pellets and the clad tube. As the linear power density in the fuel rod is increased beyond a predetermined value, the above-mentioned gap is reduced to zero, so that the mechanical interaction takes place between the clad tube and the fuel pellets to cause a stress and strain in the clad tube resulting, in some cases, a failure of the clad tube.
Various countermeasures have been taken up to now for preventing generation of excessive stress and strain in the clad tube of the fuel rod when the output power of the nuclear reactor is being increased. For instance, it has been proposed and actually adopted to control the operation of the control rods such that the output of the fuel rod does not exceed a predetermined threshold linear power. It has been attempted also to start up the nuclear reactor in such a way that, when a predetermined linear power of a fuel rod is exceeded, the flow rate of the coolant is controlled to increase the output power of the nuclear reactor at such a rate of increase of linear power as not to exceed a predetermined critical rate.
These known measures, however, encounter various problems as follows. For instance, there have been no means for offering information as to the level of the linear power of a fuel rod at which the failure of the fuel rod is caused during operation of a nuclear reactor, particularly in the course of the start up thereof. In addition, since the limit value of the rate of increase of linear power is selected to have a sufficient margine to ensure the safety, so that many days are required for starting up the nuclear reactor thereby to lower the rate of operation of the nuclear reactor.
In the starting up of a nuclear reactor, it is desirable to calculate the failure even as a probability in the fuel rod for the expected increment of the output power, and to determine the path of increase of the output power such that the failure even as a probability does not exceed a predetermined value. This starting method is preferred because it can ensure the behavior of the fuel rod at the safe side while maximizing the increment of the output power to permit an efficient operation of the nuclear reactor.
It has been proposed to use a POSHO (power shock) model in the calculation of the failure event probability. The POSHO model, which is referrred to also as EPRI model, is outlines in EPRIL NP-409.
This POSHO model is for calculating the failure event probability in fuel rods, through the evaluation of possibility of pellet-clad interaction (reeferred to as "PCI", hereinunder), taking into account various factors such as the thermal expansion, creep, welling and relocation of the fuel pellets, as well as thermal expansion and creep of the clad tube.
The determination of the failure event probability by POSHO model, however, is still unsatisfactory in that it takes into account only the case where the stress exceeding a threshold value is applied to the clad tube, in the evaluation of possibility of PCI. Namely, with this method, the nuclear reactor has to be operated at a power increment which is sufficiently smaller than the value obtained from the failure event probability to ensure safety, so that many days are required for the start up of the nuclear reactor.